ASTM International - ASTM E900-87(1994)

Standard Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials, E 706 (IIF)

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Organization: ASTM International
Publication Date: 9 July 1987
Status: historical
Page Count: 5
scope:

1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 41-J (30-ftlbf) data. Radiation damage calculative procedures have been developed from a statistical analysis of an irradiated material data base that was available as of June 1982, and checked against readily available data up to August 1983. In the procedure, a chemistry factor given in tabular form as a function of copper and nickel contents, is multiplied by a fluence factor read from a graph or calculated from a formula. A difference between this guide and the earlier edition is the addition of nickel content in the chemistry factor. This guide is applicable for the following specific materials, range of irradiation temperature, neutron fluence, and fluence rate:

1.1.1 Materials

1.1.1.1 A 533 Type B Class 1 and 2, A302 Grade B, A302 Grade B (modified), A508 Class 2 and 3.

1.1.1.2 Submerged arc welds, shielded arc welds, and electroslag welds for materials in 1.1.1.1.

1.1.1.3 Weld heat-affected zones of the materials in 1.1.1.1 and 1.1.1.2.

1.1.2 Copper contents within the range from 0.01 to 0.40 weight %.

1.1.3 Nickel content within the range from 0 to 1.2 weight %.

1.1.4 Irradiation exposure temperature within the range from 530 to 590°F (277 to 310°C).

1.1.5 Neutron fluence within the range from 1 by 10 17 to 1 by 1020 n/cm2 (E > 1 MeV).

1.1.6 Neutron fluence rate and energy spectra within the range expected at the reactor vessel core beltline region of light-water cooled reactors.

1.2 The basis for the method of adjusting the reference temperature is a report describing the basis for Regulatory Guide 1.99. The report is based on the reactor vessel surveillance data and analyses described by Guthrie and Odetle and Lombrozo; the extent of that data base is indicated by the dashed lines in .

1.3 This guide is Part IIF of Master Matrix E 706 which coordinates several standards used for irradiation surveillance of light-water reactor vessel materials. Methods of determining the applicable fluence for use in this guide are addressed in Master Matrix E 706, Practices E 560 (IC) and E944 (IIA), and Method E 1005 (IIIA). The overall application of these separate guides and practices is described in Practice E 853 (IA).

1.4 The values given in inch-pound units are to be regarded as the standard. The values given in parentheses are for information only.

1.5 This standard does not purport to address all of the safety concerns, if any, associated with its use. It is the responsibility of the user of this standard to establish appropriate safety and health practices and determine the applicability of regulatory limitations prior to use.

Document History

February 1, 2015
Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
1.1 This guide presents a method for predicting values of reference transition temperature shift (TTS) for irradiated pressure vessel materials. The method is based on the TTS exhibited by Charpy...
February 1, 2015
Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials
1.1 This guide presents a method for predicting values of reference transition temperature shift (TTS) for irradiated pressure vessel materials. The method is based on the TTS exhibited by Charpy...
July 15, 2007
Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)
1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf...
June 10, 2002
Standard Guide for Predicting Radiation-Induced Transition Temperature Shift in Reactor Vessel Materials, E706 (IIF)
1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 30-ftlbf...
ASTM E900-87(1994)
July 9, 1987
Standard Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials, E 706 (IIF)
1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 41-J...
July 9, 1987
Standard Guide for Predicting Neutron Radiation Damage to Reactor Vessel Materials, E 706 (IIF)
1.1 This guide presents a method for predicting reference transition temperature adjustments for irradiated light-water cooled power reactor pressure vessel materials based on Charpy V-notch 41-J...
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